RBMK coupled neutronics/thermal-hydraulics analyses by two independent code systems

Conference · Sat Jul 01 00:00:00 EDT 2006

OSTI ID:22039662

Parisi, C; D'Auria, F ^[1]; Malofeev, V ^[2]; Ivanov, B; Ivanov, K ^[3]

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1. Univ. of Pisa, Dept. of Mechanical, Nuclear and Production Engineering, via Diotisalvi, 2, 56100 Pisa (Italy)
2. Kurchatov Inst., Kurchatov Square 1, Moscow 123182 (Russian Federation)
3. Pennsylvania State Univ., RDFMG, 230 Reber Building, Univ. Park, PA 16802 (United States)

This paper presents the coupled neutronics/thermal-hydraulics activities carried out in the framework of the part B of the TACIS project R2.03/97, 'Software development for accident analysis of RBMK reactors in Russia'. Two independent code systems were assembled, one from the Russian side and the other from the Western side, for studying RBMK core transients. The Russian code system relies on the use of code UNK for neutron data libraries generation and the three-dimensional neutron kinetics thermal-hydraulics coupled codes BARS-KORSAR for plant transient analyses. The Western code system is instead based on the lattice physics code HELIOS and on the RELAP5-3D C code. Several activities were performed for testing code system's capabilities: the neutron data libraries were calculated and verified by precise Monte Carlo calculations, the coupled codes' steady state results were compared with plant detectors' data, and calculations of several transients were compared. Finally, both code systems proved to have all the capabilities for addressing reliable safety analyses of RBMK reactors. (authors)

• Research Organization: American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
• OSTI ID: 22039662
• Resource Relation: Conference: PHYSOR-2006: American Nuclear Society's Topical Meeting on Reactor Physics - Advances in Nuclear Analysis and Simulation, Vancouver, BC (Canada), 10-14 Sep 2006; Other Information: Country of input: France
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
97 MATHEMATICAL METHODS AND COMPUTING
ACCIDENTS
B CODES
H CODES
LWGR TYPE REACTORS
MONTE CARLO METHOD
NUCLEAR DATA COLLECTIONS
R CODES
REACTOR KINETICS
REACTOR LATTICE PARAMETERS
SAFETY ANALYSIS
STEADY-STATE CONDITIONS
THERMAL HYDRAULICS
THREE-DIMENSIONAL CALCULATIONS